ML063110234
| ML063110234 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 11/03/2006 |
| From: | Chamberlain D Division of Reactor Safety IV |
| To: | James Drake, Fairbanks A, David Loveless Division of Reactor Safety IV |
| References | |
| Download: ML063110234 (4) | |
Text
November 3, 2006 MEMORANDUM TO: David Loveless, Senior Reactor Analyst Division of Reactor Safety Jim Drake, Operations Engineer Division of Reactor Safety Abin Fairbanks, Reactor Inspector Division of Reactor Safety FROM: Dwight D. Chamberlain, Director /RA/
Division of Reactor Safety
SUBJECT:
SPECIAL INSPECTION CHARTER TO EVALUATE THE LOSS OF FEEDWATER AND SUBSEQUENT REACTOR TRIP AT RIVER BEND STATION.
On October 19, 2006, an inadvertent main feedwater isolation occurred when a chart recorder mechanism was dropped onto the main control panel. The loss of feedwater resulted in a low reactor vessel water level reactor trip followed by an unnecessary main steam isolation. Based on the results of an evaluation conducted in accordance with Management Directive 8.3, "NRC Incident Investigation Program," a special inspection will be performed to inspect the circumstances surrounding this event and the licensee's actions in response to the event. You are hereby designated as the Special Inspection Team members. David P. Loveless is designated as the team leader.
A. Basis On October 19, 2006, at 1756 CDT, River Bend Station experienced a reactor scram from 100 percent power because of a loss of all feedwater followed by the closure of main steam isolation valves. Following the scram, operators controlled reactor coolant system pressure by manually cycling safety relief valves. Given that the reactor core isolation cooling system was tagged out at the time of the event, operators controlled reactor pressure vessel level with the high pressure core spray system by overriding the high pressure core system injection valve and cycling it open and closed. Safety relief valve pressure control resulted in emergency operating plan entry conditions for containment pressure and suppression pool level.
The cause of the scram was a loss of feedwater caused when a chart paper mechanism from a chart recorder located on the control panel was inadvertently dropped onto the closure switches for the outboard feedwater isolation valves. An operator had pulled the chart recorder out of the panel to correct a problem with the chart paper. With the chart
David Loveless recorder withdrawn, the chart paper mechanism fell off the recorder and onto the control panel, striking the closed pushbuttons for the valves. As a result of the long stroke time for these valves, the operator did not notice that the valves were closing and continued on with his duties. The feedwater isolation resulted in a reactor water Level 3, which generated the reactor scram. A third valve was affected by the dropped chart paper mechanism. Feedwater long cycle cleanup isolation Valve FWS-103 opened 10-20 seconds before dual indication was received on the outboard feedwater isolation valves. This did not have any adverse impact since a second valve in the line was closed.
Reactor pressure vessel level continued to decrease, and at Level 2 high pressure core spray initiated automatically and recovered water level. The reactor core isolation cooling system was tagged out for maintenance at the time of the event. At Level 2, the recirculation pumps tripped and containment isolation valves in multiple systems actuated.
Approximately 4 minutes after the scram, the main steam isolation valves closed on low steam header pressure at approximately 849 psig. The main steam isolation valve closure would normally be bypassed following a scram by operators placing the Mode switch in shutdown, however, the Mode switch was incorrectly left in "Run." With the main steam isolation valves shut, operators controlled reactor pressure with the safety relief valves.
Operators restored the feedwater lines and opened main steam isolation valves to establish normal reactor level and pressure control.
This special inspection is being chartered to review the plant and operator response to the event.
B. Scope The team is expected to address the following:
- 1. Develop a complete sequence of events, including plant response, operator actions, and the timing of any barriers that should have prevented performance errors identified.
- 2. Evaluate plant response to the conditions that existed. Determine if the plant responded as expected.
- 3. Evaluate operator response to the event. This should include the failure to verify status of plant equipment after the chart paper mechanism was dropped and the failure to place the Mode switch in shutdown. Additionally, evaluate reactor pressure vessel water level and pressure response, which manually controlled using the high pressure core spray system and the safety-relief valves.
- 4. Evaluate main control room command and control during the event.
David Loveless 5. Evaluate operator training, standards and expectation for response to abnormal conditions and how these related to the subject event.
- 6. Evaluate licensee's root cause and corrective actions, including oversight by the onsite review committee.
- 7. Licensees root cause and corrective actions, including an evaluation of the over lap of the post-trip review discovery and the timing of the onsite review committee meeting.
C. Guidance Inspection Procedure 93812, "Special Inspection," provides additional guidance to be used by the Special Inspection Team. Your duties will be as described in Inspection Procedure 93812. The inspection should emphasize fact-finding in its review of the circumstances surrounding the event. It is not the responsibility of the team to examine the regulatory process. Safety concerns identified that are not directly related to the event should be reported to the Region IV office for appropriate action.
The Team will report to the site, conduct an entrance, and begin inspection no later than November 7, 2006. While onsite, you will provide daily status briefings to Region IV management, who will coordinate with the Office of Nuclear Reactor Regulation, to ensure that all other parties are kept informed. A report documenting the results of the inspection should be issued within 30 days of the completion of the inspection.
This Charter may be modified should the team develop significant new information that warrants review. Should you have any questions concerning this Charter, contact me at (817) 860-8180.
cc via E-mail:
B. Mallett T. Gwynn B. Vaidya K. Kennedy A. Howell A. Vegel R. Caniano V. Gaddy V. Dricks W. Maier P. Alter M. Miller W. Walker J. Lamb
David Loveless SUNSI Review Completed: __Y______ ADAMS: / YesG No Initials: __DPL____
/ Publicly Available G Non-Publicly Available G Sensitive / Non-Sensitive S:\DRS\_Special Inspections\RBcharterNovember2006_1.wpd SRA C:PBC D:DRS DLoveless/lmb KMKennedy DDChamberlain
/RA/ /RA/ /RA/
11/3/06 11/3/06 11/3/06 OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax